Search

In order to understand the recycling and emission processes of deuterium atoms, spectral profiles of the line emitted from the divertor region of JT-60U have been observed with a high-resolution spectrometer and analysed by simulation with a three-dimensional neutral particle transport code. The profile has been explained as composed of narrow and broad components; the narrow component is attributed to dissociative excitation and electron collisional excitation of the atoms produced by dissociation, and the broad component is attributed to electron collisional excitation of the atoms produced by reflection and charge exchange. In low-density plasmas, the simulated line profile agrees reasonably well with that observed, although the component attributed to the atoms reflected at the divertor tiles is overestimated by a factor of about two. Dissociative excitation from deuterium molecules and molecular ions plays an important role for the line intensity. The ratio of the line intensity to the deuterium atom flux for high-energy deuterium atoms, which are produced by the reflection and charge exchange, is reduced, because the fast atoms readily escape from the divertor plasma. The width of the narrow component in a low-density case corresponds to a temperature of deuterium atoms of 1.3 eV, and that in a high-density case corresponds to a temperature of 2.2 eV.

In JT-60U, a visible spectroscopic system with high wavelength resolution has been developed to observe Doppler broadening of spectral lines emitted from divertor plasmas. An echelle grating is used to obtain the high wavelength resolution and the nominal wavelength resolution is 5.3 pm. Using optical fibre arrays, the Littrow mounting and a charge-coupled device camera, it is possible to observe 10 sight lines simultaneously with a spatial resolution of 1.2 cm.

A motional Stark effect polarimeter with five viewing points has been installed to obtain current profiles in the JT-60U tokamak. Offset angles caused by Faraday rotation were calibrated by rapidly moving a plasma, keeping the q profile unchanged. Polarized background light was observed and its effect was corrected by detecting the light which had a wavelength close to that of the beam emission. Current profiles in a current rampdown discharge were measured, and the q value at the inversion radius of the sawtooth oscillation agreed with the theoretical value (unity) within 10%.

CD/CH-band intensities emitted from hydrocarbon molecules have been measured in the divertor region of JT-60U and the chemical sputtering yield of methane was estimated as a function of the surface temperature and the deuterium ion flux. The chemical sputtering yield increases with the surface temperature and decreases with increasing ion flux density in the L-mode plasmas. The B4C converted CFC tiles are installed in JT-60U and it is found that the chemical sputtering of B4C converted CFC tiles is suppressed in comparison to normal CFC tiles.

In order to measure the line integrated electron density on the International Thermonuclear Experimental Reactor (ITER), a vibration-compensated five-channel CO2 laser interferometer/polarimeter system is being developed. The design of the interferometer/polarimeter has been improved from the previous design, which used a CO2 laser (wavelength, λ=10.6 μm) and a CO laser (λ=5.3 μm), on the basis of experience gained with a dual CO2 system (λ=10.6 and 9.27 μm) on JT−60U. The JT−60U system gives good reliability for both interferometry and polarimetry operation. The performance of the dual CO2 laser is expected to meet the requirements for measurements on ITER (accuracy; 1%, time resolution; 1 ms).

Spatial distributions of absolute line intensities of D I and C II-IV in the divertor region have been measured simultaneously with a 38-channel optical fibre array in JT-60U. The radiative losses in the divertor region have been derived from the intensities. In L-mode discharges without the MARFE, the fractions of the radiative losses due to deuterium and carbon are 36% and 64%, respectively. The carbon radiation is related to the carbon influx by a simple impurity model based on the confinement time of the carbon ion on the assumption of the physical sputtering. The power radiated due to carbon accounts for a large increase in the radiative losses by the MARFE.

A two-dimensional impurity code based on Monte Carlo techniques (IMPMC) has been developed, in order to study the impurity behavior in the divertor plasma. The model includes (1) impurity generation at the divertor plate, (2) ionization of sputtered neutrals, (3) parallel motion of impurity ions along field lines, (4) Coulomb scattering, (5) cross-field diffusion, and (6) atomic processes. This model is applied to the carbon impurity behavior in a JT-60U NB heated plasma. The impurity generation mechanism, the effect of the thermal force on impurity sheilding, and the contribution of charge exchange recombination to C 3+ ion density near the plates are clarified. The measured spatial profile of the CIV line indicates that the diffusion coefficient, D ⊥ is around 1 m 2 /s. The ion temperature measured by Doppler broadening of the CIV line is found to be shifted to lower temperature side due to the overspread distribution of C 3+ ions in the SOL by the thermal force.

B 4 C conversion-coated carbon fiber composites (B 4 C-converted CFC) have been installed on the outboard strike point of the divertor plate in JT-60U to form a complete toroidal ring. The plasma-surface interactions of the B 4 C-converted CFC have been investigated in high power divertor operation with neutral beam (NB) heating power of - 30 MW for 2 s. It is found that the carbon impurity is reduced without a significant increase of the boron impurity using a 100 μm thick B 4 C layer with a B /C ratio of 2.7. The oxygen impurity is reduced significantly by the evaporation of a 300 μm thick B 4 C layer with a B /C ratio of 3.8. A reduction in the detrimental effects on plasma-surface interactions with carbon-based materials has been demonstrated for more than 1000 shots by the use of surface-boronized carbon tiles in JT-60U

In JT-60U boronization using decaborane was carried out. Boronization reduced the oxygen impurity in OH discharges and shortened the wall conditioning after the vacuum vessel vent and consequently enabled JT-60U to produce clean plasmas easily except for NB heated plasmas. After boronization, particle recycling was reduced drastically in OH and NB discharges. High confinement plasmas were obtained including high β p mode and H-mode discharges. In the latest boronization part of divertor plates were replaced with B 4 C coated tiles with a B 4 C thickness ∼ 300 μm. After introducing B 4 C divertor tiles, an explosive generation of boron particles from the tiles was observed. By the combined effects of boronization with decaborane and boron generation from B 4 C tiles, oxygen impurity was so low that oxygen line signals were reduced to noise levels after the latest boronization. On the other hand, boron burst from the B 4 C tiles restricted the operation of JT-60U.

Helium (He) transport and behavior have been investigated during ELMy (edge-localized mode) H-mode using a helium beam in JT-60U. Helium transport coefficients in the ELMy H-mode plasmas ( H = 1.5) are found to be D A = 1–2 m 2 /s and C V = 1.0–1.5 when a neutral beam of 60 keV helium atoms is used for a source of central fuelling, and D A = 1 m 2 /s and C V = 1.0–1.5 when a short pulse He gas puff is used as peripheral fuelling. No helium accumulation is observed during ELMy H-mode. The behavior of helium in the divertor region has been studied with divertor spectroscopic measurement and compared with that in the main plasma. The He recycling flux in the divertor during ELMy H-mode is about one order of magnitude less than during L-mode with the same level of He gas fuelling.

The performance of two array antennae of phased 2 × 2 loops for JT-60U is presented in the initial operation for higher harmonic ion cyclotron resonant frequency (ICRF) experiments. The array antenna is designed so as to have a sharp coupled power spectrum of n  peak ≈ 3 with Δn  ≈ 2.5 when out of phase with toroidally adjacent loop currents ((π, 0) phasing) at 116 MHz by separating the loops widely by 0.44 m. The maximum coupled power up to the present is 6.4 MW with (π, 0) phasing at Bt = 4 T, which corresponds to the power density of about 6.1 MW m−2. Anomalous antenna breakdown is observed during combined neutral beam injection (NBI) and ICRF heating for the particular magnetic field strength Bt = 3.0 T at which an ion cyclotron resonance layer of 7ωcD is located near the antennae. A fairly large antenna loading resistance is obtained, i.e. 4 Ω for (π, 0) phasing at 0.13 m of the distance δSF between the separatrix and the Faraday shield, and 5 Ω for in-phase ((0, 0) phasing) at very large distances of more than 0.33 m. results show the decrease in the loading resistance at the L to H transition is about 30%, from 5 to 3.7 Ω with (0, 0) phasing at δSF = 0.32 m. The ratio of the incremental radiation loss to ICRF power decreases with increasing ICRF power and is saturated at small values less than 10% at power levels more than 4 MW. The increase in the radiation loss seems to be mainly caused by the light impurities from the first wall.

Ion temperature profiles of neutral beam heated plasmas in JT-60 and TFTR have been simulated using models of ion thermal diffusivity, χi, based on ion temperature gradient mode (ηi mode) turbulence and drift wave turbulence (trapped electron mode and circulating electron mode). The ion temperature profiles measured by charge exchange recombination spectroscopy are compared to predicted Ti profiles calculated from three theoretical models for ion heat transport by Dominguez and Waltz (1987), Lee and Diamond (1986), and Romanelli (1989). All three χi models can reproduce the measured Ti profile over a wide range of parameters in JT-60 L mode plasmas, except for two 1 MA limiter cases. With the use of transport models, which were adjusted to fit JT-60 plasmas, it was found that the Dominguez and Waltz and Romanelli models agree with measured Ti, profiles in TFTR L mode discharges in the region a/4 < r < a/2. The observed central peaking of Ti near the magnetic axis of TFTR L mode plasmas cannot be reproduced. It was found that the Lee and Diamond model does not fit me data as well. The χi models studied cannot consistently reproduce the measured Ti in the peripheral region of either JT-60 or TFTR plasmas. In the high ion temperature (high Ti) plasmas of JT-60 and supershot plasmas of TFTR, the predicted Ti profiles are much broader than the measured profile. In supershot plasmas, the measured central ion temperature greatly exceeds the predicted temperature, although there is reasonable agreement near the edge

In JT-60U, spectra of deuterium and impurities in divertor plasmas were observed with absolutely calibrated VUV spectrometers. Charge state distributions for C II-IV and O III-VI were obtained from the spectral line intensities. The Li-like ion was dominant for carbon and the B- and Be-like ions were dominant for oxygen. Radiative power losses from divertor plasmas were calculated from the spectroscopic measurements. For a neutral beam heated discharge with a heating power of 12 MW, the fractions of the radiative power losses due to deuterium, carbon and oxygen were 58, 39 and 3.4%, respectively

In JT-60, hot ion modes (Ti(0) = 9-13 keV, H0 beam into H+ plasma) exhibiting peaked profiles of the ion temperature, Ti(r), and the toroidal rotation velocity, Vt(r) have been obtained with an effective safety factor qeff ~ 10. Experiments have been performed to study the characteristics of these peaked profiles in which qeff was scanned over the range 4.5 < qeff < 15. Peaking of the profiles of Ti(r) and Vt(r) occurs within a limited region. Moreover, the boundary of the region of improved performance is well characterized by the q = 1, 2 and 3 surfaces. The improvement in global confinement in JT-60 hot ion modes is attributed to the substantial increase in the volume of the enhanced confinement region within the q = 3 surface, in which region the peaked Ti profile is generated

The impurity behavior and the remote radiative cooling have been studied in the JT-60U divertor discharges with high power NB heating. NB at a power level of 20 MW was injected for 2 s, and no carbon bloom has been observed. For the first time, the measurement and the calculation for the carbon influx from the divertor plates are compared quantitatively. The carbon generation mechanism can be explained by the sputtering with deuterium (physical sputtering), oxygen and carbon. It is found that the contribution of the chemical sputtering with deuterium is small and that of the sputtering with oxygen is important. Remote radiative cooling can reduce the heat flux onto divertor plates. The dependence of the radiative cooling power on the electron density and the safety factor is presented. The favorable operation regime for discharges with high power auxiliary heating is discussed based on the experimental results.

The effects of wall conditioning techniques used in JT-60U are described. Helium glow discharge cleaning is effective in reducing oxygen and particle recycling. Both carbon and oxygen impurities are reduced by a factor of 2 by reducing the wall temperature from 300 to 150°C. The dilution ratio H/(H+D) decreased by D 2 -TDC, ohmic discharges and He-GDC. At the initial phase of wall conditioning, the dilution ratio increased by He-TDC, and a sequence of tokamak discharges with neutral beam injection, suggesting that hydrogen particles were sputtered from inside the graphite wall. After sufficient wall conditioning had reduced the dilution ratio below 0.25, the dilution rationdid not deteriorate by neutral beam injection or by a sequence of shots.

In JT-60U, impurity profiles were measured with charge exchange recombination spectroscopy (CXRS) and impurity transport was analyzed. Density profiles of fully ionized carbon ions C6+ were obtained from neutral beam density profiles and excitation rate coefficients for CVI 529.0 nm (n=8-7). Peaked density profiles of C6+ in a typical NB heated deuterium plasma were measured. Peaking parameter Cν was not constant and was in the range of 0.5–1.0. OVII (1s2-2enp, 2.16 nm) and OVIII (1s-2p, 1.90 nm) were measured with a grazing incidence spectrometer viewing the plasma center vertically. Oxygen impurity transport was analyzed in the peripheral region from the ratio of OVIII/OVII.

Improvement of both energy and particle confinement was observed in beam heated hydrogen plasmas in JT-60 when the heating power and the line averaged density exceeded thresholds. The radiation power from the divertor plasma increased to as much as 50% of the heating power. This regime was termed improved divertor confinement (IDC). The discharges with improvement of the energy confinement by up to 20% and enhanced divertor radiation lasted for several seconds in a quasi-steady state without significant impurity accumulation in the core. The threshold line averaged electron density for the IDC regime increased linearly with power and dropped with increasing safety factor. The improvement was observed only when the ion Del B drift was towards the X-point. The behaviour of light impurities and in-out asymmetries in particle recycling changed with reversal of the toroidal magnetic field. These findings are consistent with the theoretical model in which collisional transport near the separatrix is connected with the onset of the IDC regime

The first experiment simulating helium ash exhaust in a DT fusion reactor has been performed in the JT-60 tokamak with {ital L}-mode divertor discharges. He was injected with a 30-keV neutral beam to realize core fueling. Divertor neutral pressures of hydrogen and He increased in proportion to the cube of the electron density of the main plasma. The He enrichment factor was 0.5 at {ital {bar n}}{sub {ital e}}=6{times}10{sup 19} m{sup {minus}3} and increased in proportion to {ital n}{sub {ital e}}. In an {ital L}-mode plasma of a fusion reactor, thermalized {alpha} particles (He ash) generated in the core plasma can be easily exhausted with a pumping speed of several tens m{sup 3}/s.

A diagnostic system using a helium beam for local poloidal magnetic field measurements (Zeeman polarimeter system) has been constructed. The Zeeman spectrum of the helium 31P‐21S line was observed in JT‐60. The intensity and energy of the neutral helium beam were up to 0.6 A and 200 keV, respectively. Resolution of the Zeeman polarimeter was 0.01 nm which was realized by using a Littrow‐type spectrometer. Photon counting efficiency was 0.065. Moreover, an in situ wavelength calibration could be done, using an Ar laser, within an error of 0.001 nm.

Results are presented of the second harmonic heating experiments in the ion cyclotron range of frequencies (ICRF) on JT-60. Heating and confinement characteristics are investigated by using the hydrogen second harmonic (H-majority 2ωcH scheme and the hydrogen minority second harmonic (H-minority 2ωcH scheme (nH/nHe ≈0.1) in two toroidal phasing modes of antenna currents, in phase ((0,0)) and out of phase ((π, 0)), at PIC ≤ 3 MW, ne =(1.3−6.6) × 1019 m−3 and Ip = 1–2.4 MA. Efficient plasma heating has been demonstrated in both heating schemes. Intense non-Maxwellian ion tails are observed at low density. Sawtooth oscillations with an enhanced period in the electron temperature are found over a wide range of electron densities. The incremental energy confinement time, τEinc, for (π, 0) phasing is larger than for (0,0) phasing, irrespective of the heating scheme. In particular, H-minority 2ωcH heating with (π, 0) phasing shows excellent plasma heating with τEinc ≈ 100–120 ms, which is larger than that of neutral beam injection (NBI) heating by a factor of about two. Different dependences of τEinc on Ip are obtained in the two heating schemes. The (0,0) phasing data of τEinc in H-majority 2ωcH heating increase with Ip up to 1.9 MA, while the (π, 0) phasing data in H-minority 2ωcH heating are almost constant with Ip up to 2.4 MA. The global energy confinement shows an L-mode behaviour. The energy confinement time for H-minority 2ωcH heating at high density agrees well with the Shimomura-Odajima scaling.

The detailed design of the Impurity Influx Monitor (divertor) has been carried out to provide the measurement capability in the harsh environment such as higher irradiation levels of neutron, gamma‐ray and particles than in present devices. The in‐situ calibration system using a micro retro‐reflector array has been developed to monitor the sensitivity change of the optical system due to the environmental effects. The optical alignment system for the Monitor has been developed by using a dedicated optics for alignment in the collection optics for measurement.

Helium ash exhaust characteristics have been studied in the JT-60 lower divertor with ohmic heating (OH) and neutral beam heating (NBH) of PNB ≦ 18 MW and Ip = 1 MA. Helium fueling was done by gas puff. The divertor neutral pressures of hydrogen (H2) and helium were measured by a residual gas mass analyzer. The helium enrichment in the divertor was also measured. In OH discharges with helium gas puff, no buildup of helium neutral pressure was observed and helium was depleted. On the other hand, in NBH discharges with helium gas puff, although enrichment of helium is not observed, the neutral pressures of H2 and helium increase in proportion to . The density dependence of the helium neutral pressures in NBH discharges is promising for efficient helium ash exhaust in fusion reactors.

In JT-60, MARFEs have been observed frequently in high- I p and high density limited discharges with neutral beam heating after replacement of the first wall from TiC coated molybdenum tiles to graphite ones. The MARFEs occur near the density limit of the JT-60 plasmas. The NB heating is effective to suppress the MARFE. The empirical scaling of the threshold density for MARFE onset taking account of the NB power is obtained, which is useful to predict the MARFE onset. The radiated power from the plasma with MARFE is about 90% of the absorbed power. Both stored energy and central electron temperatures do not change with the MARFE onset in spite of the such intense radiation loss. So the MARFE does not affect the core plasma. A simple model based on the thermal equilibrium in a peripheral plasma can explain the onset condition of the MARFE.

Module-type flat-field grazing-incidence spectrographs with holographic gratings and multichannel detectors for large TOKAMAK (JT-60) plasma diagnosis are developed. The spectrographs cover the different wavelength regions from 0.5–122 nm, and are set to measure impurity lines in the plasma every 20 ms with space resolution of 7 cm. The flat-field imaging properties with designed wavelength resolution were confirmed, and results of tokamak plasma measurements proved the value of these spectrographs for plasma diagnosis.

Shine-through for hydrogen and helium beams in a plasma was measured. Shine-through for an atomic hydrogen beam with f1:f2:f3=7%:32%:61% (where f1, f2 and f3 represent fractions of the full, half and third energy, respectively) and a full energy of 140 keV, was smaller than the value predicted by usual cross-sections without inclusion of excitation processes at an electron density (ne) above 3*1013 cm-3. A fractional beam stopping increment ( delta ), which was defined by delta =( sigma s- sigma s(0)) sigma s(0) (here sigma s was an experimental stopping cross-section and sigma s(0) was the cross-section without including multistep processes via excited states), was about 0.2 at ne=(3.1-5.0)*1013 cm-3 and Zeff=1.6-2.5. The value of delta was consistent with the calculation by Janev et al. (1989, Nucl. Fusion 29, 2125). For a 160-190 keV helium beam, the enhancement of the stopping cross-section was not observed, as expected.

In JT-60, H-mode experiments with outer and lower divertors have been performed. In the outer divertor discharge, an H-mode similar to the modes observed in the lower/upper divertor discharges is obtained. Its threshold absorbed power and electron density are 16 MW and 1.8 × 1019 m−3. In the two schemes of combined heating with NB + ICRF and NB + LHRF, H-mode discharges are also obtained. Moreover, in the new configuration with the JT-60 lower divertor, H-mode phases with and without edge localized modes are obtained. The improvement in the energy confinement time in both divertor configurations is limited to values within 10%. The paper mainly presents the H-mode results of the outer divertor discharges. Also, typical results of the lower divertor discharges are shown for a comparison of the H-mode characteristics of the two configurations.

The effects of a radial electric field (EF) on the losses of alpha particles and neutral beam injection (NBI) fast ions in typical ITER operation scenarios for both error fields due to test blanket modules (TBMs) and toroidal field (TF) ripple were evaluated using an iterative method to execute an orbit-following Monte-Carlo code and a one-dimensional transport code. The EF effect on the loss of fast ions strongly depends on the operation scenario as well as on the error field. The electric field is very significant in the loss of fast ions in a 9 MA ITER operation scenario with a higher safety factor and in the error field associated with TBMs. The EF effect in the error field of TF ripple is very small in any operation scenario. The electric field is much more significant for the loss of NBI fast ions than for that of alpha particles. The radial electric field changes the toroidal precession of fast ions and consequently alters their condition of resonance with the error field, which may account for the EF effect on the loss of fast ions in ITER with TBMs.

Many windows will be required on ITER for optical, infrared, and microwave diagnostics. The diagnostic transmission systems will be incorporate labyrinths in port shielding blocks and so the windows will not be located in high levels of neutron and gamma radiation. Nevertheless it is necessary for the window seals to be able to withstand pressure transients, temperature excursions and mechanical shocks which are significantly different from those experienced on present-day tokamaks. The objective of the RD> 200 °C, (4) Temperature ramp rate:> 20 °C/hr, (5) Temperature gradients across windows:3∼4 points for each surface of window, (6) Mechanical acceleration of structure:> 15g, duration 10ms, 1000cycles., (7) Optical property

The function of measurements of VUV radiation emanating from the plasma in the divertor region are briefly summarized. The engineering constraints that would have to be satisfied in the implementation of VUV measuring systems for this region are identified. Two proposed configurations for the installation of a VUV spectrometer are considered and the further work that is necessary to develop these design options is identified.

The design of the divertor impurity monitoring system for ITER is now well advanced. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The expected impurities are carbon, tungsten, beryllium and copper originating from the divertor target plate and from the surface of the first wall in the main chamber. Neon and other impurity gases injected into the plasma for radiation cooling in the divertor will also be observed. The wavelength range is 200 nm to 1000 nm. This system, which is one of the most important diagnostic systems for plasma control, is included in the start-up set of ITER diagnostics1.

Infrared (IR) thermography is a useful diagnostic tool for measuring surface temperatures on divertor plates by using light emitted in the IR range. Japan plans to procure and develop divertor IR thermography for the ITER project. The system design, optical design, and evaluation of detected photons have already been conducted. All optical devices will be installed in the occupied space without interfering with other devices. The required spatial resolution of 3mm on divertor plates is satisfied by the designed optics. The gamma-ray dose rate at 106 s after shutdown of operations (shutdown dose rate) in the interspace is low (15 20 μSv/h). Two color detectors and a spectroscopic system are designed using mirrors with high reflectivity in the range 1 5 μm. From the evaluation of detected photons, sufficient photons at 3 and 5 μm can be detected at low emissivity (e = 0.2).

The effects of edge-localized mode (ELM) mitigation coils (ELM coils) on the loss of NBI-produced fast ions and fusion-produced alpha particles are investigated using an orbit following Monte Carlo code. The ELM mitigation coil field (EMC field) may cause a significant loss of fast ions produced by NBI on the order of 16.0–17.0% for a 9 MA steady-state ITER scenario. A significant transit-particle loss occurs in the case of the toroidal mode number n = 4 in which magnetic surfaces are ergodic near the plasma periphery. When the number of ELM coils in each toroidal row is nine, the main toroidal mode n = 4 is accompanied by a complementary mode n c = 5. Concerning the resonance of fast-ion trajectories, the anti-resonant surfaces of n = 4 are very close to the resonant surfaces of n c = 5 and vice versa. Since the effect of resonance on fast-ion trajectories dominates that of anti-resonance, a synergy effect of the main and complementary modes effectively enlarges the resonant regions. In a single n-mode EMC field, the resonant and anti-resonant regions are well separated. The peak heat load due to the loss of NB-produced fast ions near the upper ELM coils is as high as 1.0–1.5 MW m−2, which exceeds the allowable level in ITER. Rotation of the EMC field is essential for ITER to alleviate the local peak heat load. Most loss particles hit the inner side of the torus of the dome in the ITER divertor. The loss of alpha particles is also increased by the effect of the EMC field. The loss is still acceptably low at less than 1.0%.

The in situ calibration method for the impurity influx monitor (divertor) is experimentally examined. The total reflectance of the optical path from the focal point of the Cassegrain telescope to the first mirror is derived using a micro retroreflector array. An optical fiber with angled physical contact (APC) connectors reduces the return edge reflection. APC fibers and a multimode coupler increase the signal-to-noise ratio by about one order compared to that of triple-branched fibers and enable measurement of the wavelength dependence of the total reflectance of the optical system even after potential deterioration of mirror surfaces reduces reflectance.

The intensity of emission lines in the ITER divertor plasma is synthesized using a model of the edge and scrape-off layer plasma. The intensity of hydrogen isotopes and impurities observed with the divertor central optical system of the Impurity Influx Monitor (Divertor) is estimated. In both inner and outer divertor plasmas, ionizing and recombining plasma components are observed for C I–III, and the recombining plasma component is dominant for C IV–V. Line-of-sight integrated spectra are synthesized in the wavelength range 200–1000 nm using photon emission coefficients. In order to evaluate the carbon-impurity influx, identifying the plasma component, i.e. an ionizing or recombining plasma component, is important.

Metallic mirrors will be used as plasma-viewing elements in all optical and laser diagnostic systems in ITER. In the harsh environment of ITER, the performance of mirrors will decrease mainly because of the erosion of their surfaces and deposition of impurities. The deterioration of the optical properties of diagnostic mirrors will directly affect the entire performance of the respective ITER diagnostics, possibly leading to their shutdown. Therefore, R&D on mirrors is of crucial importance for ITER diagnostics. There is a coordinated worldwide R&D programme supervised by the Specialists Working Group on first mirrors of the International Tokamak Physics Activity, Topical Group on Diagnostics. This paper provides an overview of new results in the field of first mirrors, covering the manufacturing of ITER mirror prototypes, investigations of mitigation of deposition and mirror cleaning and the predictive modelling of the mirror performance in ITER. The current status of research on beryllium deposition—a new critical area of mirror research—is given along with an outlook for future activities.

We are developing a spectroscopic diagnostics system in ultraviolet and visible wavelength regions for monitoring ITER divertor plasmas. An equivalent-size prototype of the optical components for viewing upper port fanarray chords has been assembled as a system to measure spectral throughput, i.e., etendu. Collisional-radiative models for HeI and CIV are used to estimate the emission line intensities of helium ash and carbon impurity ions in a divertor region of a burning plasma. The estimated line intensity of CIV, λ772.6 nm, satisfies ITER requirements for the time resolution of measurement of Ti. A numerical simulation of the computerized tomographic technique for various pairs of viewing fan arrays has been applied to the divertor plasma region to reconstruct a two-dimensional distribution. The optimized pair of viewing fan arrays resolves a model distribution with a reasonable spatial resolution. We measure the reflectance of surfaces of carbon-fiber-composite and tungsten blocks, which make up the plasma-facing divertor target plates and the divertor dome. The reflectance of the surface of the tungsten divertor block is 23% at Hα (λ656.3 nm). A sandblast-processed tungsten surface effectively reduces direct reflectance; the resultant reflectance is less than 0.7%.

A collective Thomson scattering (CTS) diagnostic using carbon dioxide (CO2) laser is being developed to establish a diagnostic method of confined α-particles in burning plasmas. To realize the CTS diagnostic, a transversely excited atmospheric CO2 laser has been developed. To obtain single-mode output, which is needed for the CTS diagnostic, seed laser is injected to the cavity with an unstable resonator. Using this technique, output energy of 17 J at a repetition rate of 15 Hz has been achieved. This result gives a prospect for the CTS diagnostic on International Thermonuclear Experimental Reactor (ITER). Proof-of-principle test of the CTS diagnostic is being performed with the new laser system on JT-60U. A method to improve spatial and spectral resolutions, which is subject to be resolved for the CO2 laser CTS, is proposed using a resonance at the lower hybrid frequency. Calculation results show that the scattering power increases more than a factor of 10, and spatial and spectral resolution improve about 30% if the scattered wave vector is nearly perpendicular to the magnetic field.

The optical design of Impurity Influx Monitor (divertor) was carried out for new ITER design and ray-tracing analysis shows that the spatial resolution of ITER requirement (50 mm) will be achieved by these optical systems designed here. The mechanical design of front end optics also carried out based on the optical design and results of port integration. In the upper port, front end optics can be installed inside the pipe of inner diameter of φ300 mm. In the equatorial port, all the optical component are placed to avoid the interaction with other diagnostic equipments. Heat analysis was carried out for the optimization the cooling method of mirrors, mirror holders and mounting modules to reduce the temperature rise caused by the nuclear heating. It indicates that mirrors can be cooled by the thermal conduction using mirror holders made of a copper alloy and making many cooling channels on the mounting module. Finally, the effect of the thermal deformation to the optical properties is estimated by using the optical design code.

A high-repetition and high-energy transversely excited atmospheric carbon dioxide (CO2) laser for a collective Thomson scattering (CTS) diagnostic has been developed to establish a diagnostic method of confined α particles in burning plasmas. To excite a single-transverse and single-longitudinal mode, a continuous wave seed laser was injected to an unstable resonator with a cavity length of ∼4.4m. Pulse energy of 10J with a repetition rate of 10Hz has been achieved in the single-mode operation. Pulse energies of 18J with a repetition rate of 10Hz and 36J with single shot operation have also been achieved in the multimode operation. These results give a prospect for the CTS diagnostic on International Thermonuclear Experimental Reactor (ITER), which requires single-mode energy of 20J with repetition rate of 40Hz.

The ratio of an increment in radiation power to ion cyclotron resonance heating (ICRH) power (ΔPradmain/PIC) decreased from about 40% to below 10% with increasing ICRH power in JT-60U. The most plausible mechanism for the anomalous behavior of ΔPradmain/PICis inferred to be nonlinear wave absorption, which was experimentally estimated from the time derivative of the diamagnetic stored energy at the turn-off timing of the ICRH pulse. Nonlinear wave absorption observed during the experiments supported the theoretical estimation using a full wave code incorporating the 1-D Fokker-Planck code.

A decaborane-based boronization system has been installed in the JT-60U tokamak in order to reduce the influx of impurities during plasma discharges. Boronization has been performed under a glow discharge using a helium-decaborane gas mixture. The properties of the boron films deposited through boronization and the effects of boronization on the tokamak discharges were investigated. It was found that the deposition of a boron layer with high purity was achieved with few impurities other than hydrogen through boronization, and that the present boronization deposited toroidally nonuniform boron film. It was also found that the decaborane-based boronization resulted in good plasma performance similar to that of conventional boronization.

Plasma-surface interactions during RF heating on JFT-2 tokamak were investigated. Light impurities (O, C) and also metal impurities (Mo, Fe, Ti) increased by about twice for 220 kW RF power, however, because impurities have been reduced to an allowable level by Ti gettering onto the wall and the limiter, it was not so deleterious on plasma confinement. The increase of metal impurities can be explained by the light impurity sputtering and the effect of high energy ions produced by RF heating seems to be minor for metal release. On condition of clean surface produced by Ti gettering, decrease of electron density was observed during RF heating. This phenomenon is discussed in connection with the behaviour of high energy ion.

Impurity contamination and radiation power have been studied with a vacuum UV monochromator (10–1300 A) on JFT-2 for typical 160-kA and 18-kG hydrogen discharges. Oxygen has been found to be a dominant impurity element (5–7% of the electron density) and to be radially distributed according to the ionization sequence. The ground level population densities of the O6+ and O7+ ions in the central hot region have been determined, and that of fully ionized oxygen has been also estimated, combined with enhanced bremsstrahlung radiation measurement. The radial profile of the effective ionic charge Zeff has been determined from that of the impurity ions. The Zeff value has been found to be consistent with the value derived from the electrical conductivity. The total power of line radiation from the first resonance transitions of the oxygen ions amounts to 100–50 kW (34–17% of Joule heating power).

ICRF heating of impurity-free tokamak plasmas is studied in DIVA. The frequency is fixed at 25 MHz, which is a second-harmonic cyclotron frequency of deuteron with a toroidal magnetic field of 16.4 kG. The toroidal field BT and the ratio of proton-to-deuteron density p are varied over ranges of 12 ? BT ? 19 kG and 2 p 40%, respectively. The most favourable ion heating is obtained at BT = 18 kG and p = 5 ~ 10%. Ion temperature increases from 200 to 300 eV were measured by charge-exchange neutral analysis and also by spectroscopy, where 100 kW RF net power were applied. The results are consistent with mode-conversion theory.